Mechanical Behavior of Nuclear Reactor Materials and Components III: Ferritic Alloys II
Sponsored by: TMS Structural Materials Division, TMS: Nuclear Materials Committee, TMS: Mechanical Behavior of Materials Committee
Program Organizers: Assel Aitkaliyeva, University of Florida; Clarissa Yablinsky, Los Alamos National Laboratory; Osman Anderoglu, University of New Mexico; Eda Aydogan, Middle East Technical University; Kayla Yano, Pacific Northwest National Laboratory; Caleb Massey, Oak Ridge National Laboratory; Djamel Kaoumi, North Carolina State University
Monday 2:00 PM
March 20, 2023
Room: 28D
Location: SDCC
Session Chair: Clarissa Yablinsky, LANL; Osman Anderoglu, University of New Mexico
2:00 PM Invited
Evaluating ATF Cladding Mechanical Behavior: Benjamin Eftink1; Peter Beck1; Nan Li1; Cheng Liu1; Mathew Hayne1; Hyosim Kim1; Tarik Saleh1; 1Los Alamos National Laboratory
Nuclear fuel cladding development requires an understanding of how mechanical stresses and strains impact the cladding. There are different mechanical testing methods to extract different properties of claddings and coatings. Some of these methods include ring pull, axial tensile, bulge, and micro-cantilever. When combined they can provide a more complete picture of the claddings expected performance. In this talk, two aspects of cladding evaluation are covered for accident tolerant fuel claddings (FeCrAl and Cr coated Zircaloy), hoop direction mechanical properties and Cr coating mechanical integrity. Results will be discussed in terms of the impact of the Zircaloy/Cr coating interface on adhesion and coating cracking susceptibility in addition to considerations when measuring hoop direction properties in cladding tubes.
2:30 PM
Mechanical and Microstructural Characterization of Neutron Irradiated HT-9 Heats at LWR and Fast Reactor Relevant Temperatures: Ramprashad Prabhakaran1; Indrajit Charit2; Dan Edwards1; Mychailo Toloczko1; Stuart Maloy1; Kumar Sridharan3; 1Pacific Northwest National Laboratory; 2University of Idaho; 3University of Wisconsin-Madison
Ferritic-martensitic steels containing 9-12 weight% Cr, such as HT-9, are being considered as candidate structural materials for fast, advanced LWR, and fusion reactors due to their excellent resistance to radiation-induced void swelling, good irradiation creep properties, microstructural stability, and thermal conductivity. However, hardening, and associated embrittlement effects at irradiation temperatures below 430°C warrant further study. To further evaluate this concern, systematic investigations on the mechanical behavior and microstructure of HT-9 with slight variations in composition and heat treatment are needed over a range of doses and temperatures to better understand the sources of hardening. Mechanical and microstructural characterization were performed on three heats of HT-9 (ORNL, LANL and EBR II) with different processing conditions irradiated to 3-8 dpa in the Advanced Test Reactor at 291°C, 360°C and 431°C to understand the effects of radiation damage, and to provide insight on desirable processing conditions.
2:50 PM
The Origin of Superior IASCC Resistance of Additively Manufactured 316L Stainless Steel after Hot Isostatic Pressing in Oxygenated BWR Water: Jingfan Yang1; Laura Hawkins2; Lingfeng He2; Miao Song3; Yu Lu4; Gary Was3; Xiaoyuan Lou1; 1Purdue University; 2Idaho National Laboratory; 3University of Michigan—Ann Arbor; 4Boise State University
Recent report (M. Song, et al. Journal of Nuclear Materials 513, 33-44, 2019) showed additively manufactured (AM) 316L stainless steel (SS) after hot isostatic pressing (HIP) exhibited superior resistance to IASCC in oxygenated BWR water, as compared to 316L stainless steel in other forms. While this result caught attention from both research and regulatory communities, the mechanistic explanation to this unusual IASCC response was still not clear. Extensive investigation has been carried out to reveal the unique radiation-induced microstructural and mechanical changes that might lead to the IASCC immunity in HIP AM316L SS. The overall radiation hardening is not an accurate measure of IASCC susceptibility. A decreased strain localization along grain boundaries, caused by dislocation channel broadening, was identified as the main reason to suppress IASCC susceptibility. The root cause of channel broadening in an irradiated AM 316L SS will be discussed.
3:10 PM
Musings on Advanced Cladding Qualification: Tarik Saleh1; 1Los Alamos National Laboratory
Deploying advanced nuclear reactors will require a robust and creative framework for qualifying new cladding and core materials. The US Department of Energy, Nuclear Energy (DOE-NE) sponsored programs such as Nuclear Energy Advanced Modeling and Simulation (NEAMS) and the Fuels Cycle Research and Development - Advanced Fuels Campaign (FCRD-AFC) invest in experiments and modeling for future reactors, however a more complete strategy for Advanced Cladding Qualification (ACQ) is needed, especially as a dearth of fast test reactors limit the experimental options for relevant high dose data. This talk will present some ideas on integrating successful Advanced Fuels Qualification strategies with experimental and modeling capabilities to provide a framework for future qualification of Advanced Cladding for use in next generation reactors. While the scope of a complete ACQ campaign is vast, focused modelling and mechanical testing, along with leveraging existing data, and accelerated cladding tests, will provide paths for future qualification.
3:30 PM Break
3:50 PM
High Throughput Nanoindentation Creep Testing in Nuclear Reactor Steels: Moujhuri Sau1; Eric Hintsala2; Douglas Stauffer2; Laurent Capolungo3; Nathan Mara1; 1University of Minnesota; 2Bruker Nano Inc.; 3Los Alamos National Laboratory
Nanoindentation is a high-throughput small-scale alternative to bulk creep testing used for study nanoscale mechanical deformation over the span of seconds or minutes with minimal volumes of materials, allowing for rapid alloy testing and development. Elevated temperature tests were conducted on three different reactor steel candidate materials - an austenitic stainless steel alloy (SS347H), an ODS alloy (MA957) and a FeCrAl alloy (APMT) – to study their nanoscale mechanical response as a function of temperature and strain rate. This time-dependent deformation is quantified by calculating strain rate sensitivity, activation energy and activation volumes. The high-temperature microstructural deformation after indentation was visualized with EBSD and TEM microscopy. The data sets will be discussed in terms of dominant mechanisms responsible for creep and correlated to bulk tensile creep data.
4:10 PM
Deformation Characteristics of Additively Manufactured 316L Stainless Steels after Neutron Irradiation: Thak Sang Byun1; Maxim Gussev1; Timothy Lach1; Annabelle Le Coq1; Kory Linton1; 1Oak Ridge National Laboratory
Rapid solidification and cooling in the laser powder bed fusion process can produce a unique metastable microstructure in austenitic stainless steels (SSs), with ultrafine grains, high-density dislocations, and residual stresses. Such an additively manufactured (AM) microstructure with high defect sink density can offer desirable mechanical performance for nuclear application, including a superior combination of high strength, high ductility, and excellent radiation resistance. Neutron irradiation of miniature 316L SS tensile specimens was carried out in the high flux isotope reactor to 0.2, 2 and 8 dpa at 300 and 600 °C, and post-irradiation testing in selected conditions. This presentation is to discuss the deformation characteristics and mechanical properties of the AM SSs before and after neutron irradiation. We will elucidate unique post-irradiation plasticity behaviors observed in the austenitic structures, including the radiation-induced softening and ductilization in AM structures and serrated yielding (i.e., dynamic strain aging behavior) at high temperatures.